The present invention pertains to the fields of zirconium base alloy fuel cladding for use in pressurized water and boiling water reactors. It is especially concerned with fuel cladding having properties which minimize the adverse effects of pellet-clad interaction (PCI) in water reactor fuel elements.
The use of cladding tubes made entirely of a high zirconium alloy has been the practice in the water reactor industry. Examples of common alloys used are Zircaloy-2, and Zircaloy-4. These alloys were selected based on their nuclear properties, mechanical properties and high-temperature aqueous-corrosion resistance.
The history of the development of Zircaloy-2 and 4, and the abandonment of Zircaloy-1 and 3 is summarized in: Stanley Kass, "The Development of the Zircaloys," published in ASTM Special Technical Publication No. 368 (1964) pp. 3-27. This article is hereby incorporated by reference. Also of interest with respect to Zircaloy development are U.S. Pat. Nos. 2,772,964; 3,097,094; and 3,148,055.
It is understood that a popular zirconium base alloy in the USSR for use in water reactor applications has been Ozhennite-0.5. This alloy is believed to nominally contain: 0.2 wt.% Sn--0.1 wt.% Fe-0.1 wt.% Ni, and 0.1 wt.% Nb.
Most commercial chemistry specifications for Zircaloy-2 and 4 conform essentially with the requirements published in ASTM B350-80, (for alloy UNS No. R60802 and R60804, respectively) for example. In addition to these requirements the oxygen content for these alloys is required to be between 900 to 1600 ppm but typically is about 1200.+-.200 ppm for fuel cladding applications. ASTMB350-80 is hereby incorporated by reference.
It has been a common practice to manufacture Zircaloy cladding tubes by a fabrication process involving: hot working an ingot to an intermediate size billet or log; beta solution treating the billet; machining a hollow billet; high temperature alpha extruding the hollow billet to a hollow cylindrical extrusion; and then reducing the extrusion to substantially final size cladding through a number of cold pilger reduction passes, having an alpha recrystallization anneal prior to each pass. The cold worked, substantially final size cladding is then final annealed. This final anneal may be a stress relief anneal, partial recrystallization anneal or full recrystallization anneal. The type of final anneal provided, is selected based on the designer's specification for the mechanical properties of the fuel cladding.
One problem that has occurred in the use of fuel rods utilizing the aforementioned cladding has been the observation of cracks emanating from the interior surface of the cladding which is placed under additional stress by contact with a fractured, thermally expanding oxide fuel pellet. These cracks sometimes propagate through the wall thickness of the cladding destroying the integrity of the fuel rod and thereby allowing coolant into the rod and radioactive fission products to contaminate primary coolant circulating through the reactor core. This cracking phenomena, is generally believed to be caused by the interaction of irradiation hardening, mechanical stress and fission products, producing an environment conducive to crack initiation and propagation in zirconium alloys.
Zircaloy fuel cladding tubes having a zirconium layer bonded to their inside surface have been proposed as being resistant to the propagation of cracks initiated at the interface between the fuel pellet and cladding during water reactor operation. Examples of these proposals are provided by U.S. Pat. Nos. 4,045,288; 4,372,817; 4,200,492; and 4,390,497; and U.K. patent application No. 2,104,711A. The foregoing patents are hereby incorporated by reference.
The zirconium liners of the foregoing patents have been selected because of their resistance to PCI crack propagation without consideration of this resistance to aqueous corrosion. If the cladding should breach in the reactor, allowing coolant inside the cladding, it is expected that the aqueous corrosion resistance of the liner will be vastly inferior to that of the high zirconium alloy making up the bulk of the cladding. Under these conditions the liner would be expected to completely oxidize thereby becoming useless, relatively rapidly, while leading to increased hydride formation in the zirconium alloy portion of the cladding, thereby compromising the structural integrity of the zirconium alloy. This degradation of the cladding could lead to gross failure with significantly higher release of uranium and radioactive species to the coolant.
The art has sought to address this aqueous corrosion resistance problem by burying the zirconium layer of the aforementioned patents between layers of conventional zirconium alloys having high aqueous corrosion resistance or by substituting a dilute zirconium alloy for the internally exposed zirconium layer of the prior proposals. Examples of these designs are described in U.K. patent application No. 2,119,559. Despite these efforts there continues to be a need for water reactor fuel cladding having the excellent aqueous corrosion resistance of conventional zirconium alloys on both its inside diameter and outside diameter surfaces, while having improved PCI crack propagation resistance compared to conventional Zircaloy-2 and Zircaloy-4 fuel claddings.